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HOM Survey of the First CEBAF Upgrade Style Cavity Pair

Description: The planned upgrade of the Continuous Electron Beam Accelerator Facility (CEBAF) at the Thomas Jefferson National Accelerator Laboratory (JLab) requires ten new superconducting rf (SRF) cavity cryomodules to double the beam energy to the envisaged 12 GeV. Adequate cavity Higher Order Mode (HOM) suppression is essential to avoid multipass, multibunch beam break-up (BBU) instabilities of the recirculating beam. We report on detailed HOM surveys performed for the first two upgrade style cavities tested in a dedicated cavity pair cryomodule at 2K. The safety margin to the BBU threshold budget at 12 GeV has been assessed.
Date: May 1, 2009
Creator: Marhauser, Frank; Davis, G; Drury, Michael; Grenoble, Christiana; Hogan, John; Manus, Robert et al.
Partner: UNT Libraries Government Documents Department

Risk Informed Safety Margin Characterization Case Study: Selection of Electrical Equipment To Be Subjected to Environmental Qualification

Description: In general, the margins-based safety case helps the decision-maker manage plant margins most effectively. It tells the plant decision-maker such things as what margin is present (at the plant level, at the functional level, at the barrier level, at the component level), and where margin is thin or perhaps just degrading. If the plant is safe, it tells the decision-maker why the plant is safe and where margin needs to be maintained, and perhaps where the plant can afford to relax.
Date: September 1, 2011
Creator: Youngblood, R. & Blanchard, D.
Partner: UNT Libraries Government Documents Department

Risk Informed Safety Margin Characterization Case Study: Selection of Electrical Equipment To Be Subjected to Environmental Qualification

Description: In general, the margins-based safety case helps the decision-maker manage plant margins most effectively. It tells the plant decision-maker such things as what margin is present (at the plant level, at the functional level, at the barrier level, at the component level), and where margin is thin or perhaps just degrading. If the plant is safe, it tells the decision-maker why the plant is safe and where margin needs to be maintained, and perhaps where the plant can afford to relax.
Date: April 1, 2012
Creator: Blanchard, D. & Youngblood, R.
Partner: UNT Libraries Government Documents Department

Thermal Analysis of LWR-2 Experiment in the Advanced Test Reactor

Description: A thermal analysis was performed at the Idaho National Laboratory on the Light Water Reactor (LWR-2) experiment fuel to determine fuel temperatures and safety margins in the Advanced Test Reactor (ATR). The overall experiment description and various types of fuel proposed for this experiment is presented Reference [1]. The purpose of this paper is to present the thermal analysis for the U.S. Department of Energy (DOE) Advanced Fuel Cycle Initiative's LWR-2 irradiation experiment in the ATR.
Date: June 1, 2006
Creator: Hawkes, Grant L.
Partner: UNT Libraries Government Documents Department

Addendum to NuMI shielding assessment

Description: The original safety assessment and the Safety Envelope for the NuMI beam line corresponds to 400 kW of beam power. The Main Injector is currently capable of and approved for producing 500 kW of beam power2. However, operation of the NuMI beam line at 400 kW of power brings up the possibility of an occasional excursion above 400 kW due to better than usual tuning in one of the machines upstream of the NuMI beam line. An excursion above the DOE approved Safety Envelope will constitute a safety violation. The purpose of this addendum is to evaluate the radiological issues and modifications required to operate the NuMI beam line at 500 kW. This upgrade will allow 400 kW operations with a reasonable safety margin. Configuration of the NuMI beam line, boundaries, safety system and the methodologies used for the calculations are as described in the original NuMI SAD. While most of the calculations presented in the original shielding assessment were based on Monte Carlo simulations, which were based on the design geometries, most of the results presented in this addendum are based on the measurements conducted by the AD ES&H radiation safety group.
Date: October 1, 2007
Creator: Vaziri, Kamran
Partner: UNT Libraries Government Documents Department

INTEGRATION OF RELIABILITY WITH MECHANISTIC THERMALHYDRAULICS: REPORT ON APPROACH AND TEST PROBLEM RESULTS

Description: The Risk-Informed Safety Margin Characterization (RISMC) pathway of the Light Water Reactor Sustainability Program is developing simulation-based methods and tools for analyzing safety margin from a modern perspective. [1] There are multiple definitions of 'margin.' One class of definitions defines margin in terms of the distance between a point estimate of a given performance parameter (such as peak clad temperature), and a point-value acceptance criterion defined for that parameter (such as 2200 F). The present perspective on margin is that it relates to the probability of failure, and not just the distance between a nominal operating point and a criterion. In this work, margin is characterized through a probabilistic analysis of the 'loads' imposed on systems, structures, and components, and their 'capacity' to resist those loads without failing. Given the probabilistic load and capacity spectra, one can assess the probability that load exceeds capacity, leading to component failure. Within the project, we refer to a plot of these probabilistic spectra as 'the logo.' Refer to Figure 1 for a notional illustration. The implications of referring to 'the logo' are (1) RISMC is focused on being able to analyze loads and spectra probabilistically, and (2) calling it 'the logo' tacitly acknowledges that it is a highly simplified picture: meaningful analysis of a given component failure mode may require development of probabilistic spectra for multiple physical parameters, and in many practical cases, 'load' and 'capacity' will not vary independently.
Date: July 1, 2011
Creator: Schroeder, J. S. & Youngblood, R. W.
Partner: UNT Libraries Government Documents Department

Treatment of Passive Component Reliability in Risk-Informed Safety Margin Characterization FY 2010 Report

Description: The Risk-Informed Safety Margin Characterization (RISMC) pathway is a set of activities defined under the U.S. Department of Energy (DOE) Light Water Reactor Sustainability Program. The overarching objective of RISMC is to support plant life-extension decision-making by providing a state-of-knowledge characterization of safety margins in key systems, structures, and components (SSCs). A technical challenge at the core of this effort is to establish the conceptual and technical feasibility of analyzing safety margin in a risk-informed way, which, unlike conventionally defined deterministic margin analysis, is founded on probabilistic characterizations of SSC performance.
Date: September 1, 2010
Creator: Youngblood, Robert W
Partner: UNT Libraries Government Documents Department

A Review of Information for Managing Aging in Nuclear Power Plants

Description: Age related degradation effects in safety related systems of nuclear power plants should be managed to prevent safety margins from eroding below the acceptable limits provided in plant design bases. The Nuclear Plant Aging Research (NPAR) Pro- gram, conducted under the auspices of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, and other related aging management programs are developing technical information on managing aging. The aging management process central to these efforts consists of three key elements: 1) selecting structures, systems, and components (SSCs) in which aging should be controlled; 2) understanding the mechanisms and rates of degradation in these SSCs; and 3) managing degradation through effective inspection, surveillance, condition monitoring, trending, record keeping, mainten- ance, refurbishment, replacement, and adjustments in the operating environment and service conditions. This document concisely reviews and integrates information developed under the NPAR Program and other aging management studies and other available information related to understanding and managing age-related degradation effects and provides specific refer- ences to more comprehensive information on the same subjects.
Date: September 1, 1995
Creator: Morgan, WC & Livingston, JV
Partner: UNT Libraries Government Documents Department

MCNP5 CALCULATIONS REPLICATING ARH-600 NITRATE DATA

Description: This report serves to extend the previous document: 'MCNP Calculations Replicating ARH-600 Data' by replicating the nitrate curves found in ARH-600. This report includes the MCNP models used, the calculated critical dimension for each analyzed parameter set, and the resulting data libraries for use with the CritView code. As with the ARH-600 data, this report is not meant to replace the analysis of the fissile systems by qualified criticality personnel. The M CNP data is presented without accounting for the statistical uncertainty (although this is typically less than 0.001) or bias and, as such, the application of a reasonable safety margin is required. The data that follows pertains to the uranyl nitrate and plutonium nitrate spheres, infinite cylinders, and infinite slabs of varying isotopic composition, reflector thickness, and molarity. Each of the cases was modeled in MCNP (version 5.1.40), using the ENDF/B-VI cross section set. Given a molarity, isotopic composition, and reflector thickness, the fissile concentration and diameter (or thicknesses in the case of the slab geometries) were varied. The diameter for which k-effective equals 1.00 for a given concentration could then be calculated and graphed. These graphs are included in this report. The pages that follow describe the regions modeled, formulas for calculating the various parameters, a list of cross-sections used in the calculations, a description of the automation routine and data, and finally the data output. The data of most interest are the critical dimensions of the various systems analyzed. This is presented graphically, and in table format, in Appendix B. Appendix C provides a text listing of the same data in a format that is compatible with the CritView code. Appendices D and E provide listing of example Template files and MCNP input files (these are discussed further in Section 4). Appendix F is a complete listing ...
Date: October 25, 2011
Creator: Finfrock, S. H.
Partner: UNT Libraries Government Documents Department

Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

Description: This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.
Date: March 1, 2012
Creator: Renfro, David G; Cook, David Howard; Freels, James D; Griffin, Frederick P; Ilas, Germina; Sease, John D et al.
Partner: UNT Libraries Government Documents Department

Heartbeat Model for Component Failure in Simulation of Plant Behavior

Description: As part of the Department of Energy’s “Light Water Reactor Sustainability Program” (LWRSP), tools and methodology for risk-informed characterization of safety margin are being developed for use in supporting decision-making on plant life extension after the first license renewal. Beginning with the traditional discussion of “margin” in terms of a “load” (a physical challenge to system or component function) and a “capacity” (the capability of that system or component to accommodate the challenge), we are developing the capability to characterize realistic probabilistic load and capacity spectra, reflecting both aleatory and epistemic uncertainty in system behavior. This way of thinking about margin comports with work done in the last 10 years. However, current capabilities to model in this way are limited: it is currently possible, but difficult, to validly simulate enough time histories to support quantification in realistic problems, and the treatment of environmental influences on reliability is relatively artificial in many existing applications. The INL is working on a next-generation safety analysis capability (widely referred to as “R7”) that will enable a much better integration of reliability-related and phenomenology-related aspects of margin. In this paper, we show how to implement cumulative damage (“heartbeat”) models for component reliability that lend themselves naturally to being included as part of the phenomenology simulation. Implementation of this modeling approach relies on the way in which the phenomenology simulation implements its dynamic time step management. Within this approach, component failures influence the phenomenology, and the phenomenology influences the component failures.
Date: March 1, 2011
Creator: Youngblood, R. W.; Nourgaliev, R. R.; Kelly, D. L.; Smith, C. L. & Dinh, T-N.
Partner: UNT Libraries Government Documents Department

Ion Desorption Stability in Superconducting High Energy Physics Proton Colliders

Description: In this paper we extend our previous analysis of cold beam tube vacuum in a superconducting proton collider to include ion desorption in addition to thermal desorption and synchrotron radiation induced photodesorption. The new ion desorption terms introduce the possibility of vacuum instability. This is similar to the classical room temperature case but now modified by the inclusion of ion desorption coefficients for cryosorbed (physisorbed) molecules which can greatly exceed the coefficients for tightly bound molecules. The sojourn time concept for physisorbed H{sub 2} is generalized to include photodesorption and ion desorption as well as the usually considered thermal desorption. The ion desorption rate is density dependent and divergent so at the onset of instability the sojourn time goes to zero. Experimental data are used to evaluate the H{sub 2} sojourn time for the conditions of the Large Hadron Collider (LHC) and the situation is found to be stable. The sojourn time is dominated by photodesorption for surface density s(H{sub 2}) less than a monolayer and by thermal deposition for s(H{sub 2}) greater than a monolayer. For a few percent of a monolayer, characteristic of a beam screen, the photodesorption rate exceeds ion desorption rate by more than two orders of magnitude. The photodesorption rate corresponds to a sojourn time of approximately 100 sec. The paper next turns to the evaluation of stability margins and inclusion of gases heavier than H{sub 2} (CO, CO{sub 2} and CH{sub 4}), where ion desorption introduces coupling between molecular species. Stability conditions are worked out for a simple cold beam tube, a cold beam tube pumped from the ends and a cold beam tube with a co-axial perforated beam screen. In each case a simple inequality for stability of a single component is replaced by a determinant that must be greater than zero for ...
Date: May 29, 1995
Creator: Turner, W.C.
Partner: UNT Libraries Government Documents Department

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN C-REACTOR DISASSEMBLY BASIN

Description: C-reactor disassembly basin is being prepared for deactivation and decommissioning (D and D). D and D activities will consist primarily of immobilizing contaminated scrap components and structures in a grout-like formulation. The disassembly basin will be the first area of the C-reactor building that will be immobilized. The scrap components contain aluminum alloy materials. Any aluminum will corrode very rapidly when it comes in contact with the very alkaline grout (pH > 13), and as a result would produce hydrogen gas. To address this potential deflagration/explosion hazard, Savannah River National Laboratory (SRNL) reviewed and evaluated existing experimental and analytical studies of this issue to determine if any process constraints are necessary. The risk of accumulation of a flammable mixture of hydrogen above the surface of the water during the injection of grout into the C-reactor disassembly area is low if the assessment of the aluminum surface area is reliable. Conservative calculations estimate that there is insufficient aluminum present in the basin areas to result in significant hydrogen accumulation in this local region. The minimum safety margin (or factor) on a 60% LFL criterion for a local region of the basin (i.e., Horizontal Tube Storage) was greater than 3. Calculations also demonstrated that a flammable situation in the vapor space above the basin is unlikely. Although these calculations are conservative, there are some measures that may be taken to further minimize the risk of developing a flammable condition during grouting operations.
Date: July 12, 2011
Creator: Wiersma, B.
Partner: UNT Libraries Government Documents Department

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R- AND P-REACTOR VESSELS

Description: The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS&T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D&D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement grout (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters does not provide a significant margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel contains significantly less aluminum and thus a Portland cement grout may be considered as well. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation in the R-reactor vessel is very low for ...
Date: December 29, 2009
Creator: Wiersma, B.
Partner: UNT Libraries Government Documents Department

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS

Description: The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS&T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D&D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or portland cement grout (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the R- and P- reactor vessels. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Conservative calculations estimate that either ceramicrete or the silica fume grout may be used to safely grout the R- and P- reactor vessels. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. Grout temperatures less than 100 C should however, still provide an adequate safety margin for the ...
Date: October 29, 2009
Creator: Wiersma, B.
Partner: UNT Libraries Government Documents Department

ESTIMATED SIL LEVELS AND RISK COMPARISONS FOR RELIEF VALVES AS A FUNCTION OF TIME-IN-SERVICE

Description: Risk-based inspection methods enable estimation of the probability of spring-operated relief valves failing on demand at the United States Department of Energy's Savannah River Site (SRS) in Aiken, South Carolina. The paper illustrates an approach based on application of the Frechet and Weibull distributions to SRS and Center for Chemical Process Safety (CCPS) Process Equipment Reliability Database (PERD) proof test results. The methodology enables the estimation of ANSI/ISA-84.00.01 Safety Integrity Levels (SILs) as well as the potential change in SIL level due to modification of the maintenance schedule. Current SRS practices are reviewed and recommendations are made for extending inspection intervals. The paper compares risk-based inspection with specific SILs as maintenance intervals are adjusted. Groups of valves are identified in which maintenance times can be extended as well as different groups in which an increased safety margin may be needed.
Date: March 26, 2012
Creator: Harris, S.
Partner: UNT Libraries Government Documents Department

RISK-INFORMED SAFETY MARGIN CHARACTERIZATION

Description: The concept of safety margins has served as a fundamental principle in the design and operation of commercial nuclear power plants (NPPs). Defined as the minimum distance between a system’s “loading” and its “capacity”, plant design and operation is predicated on ensuring an adequate safety margin for safety-significant parameters (e.g., fuel cladding temperature, containment pressure, etc.) is provided over the spectrum of anticipated plant operating, transient and accident conditions. To meet the anticipated challenges associated with extending the operational lifetimes of the current fleet of operating NPPs, the United States Department of Energy (USDOE), the Idaho National Laboratory (INL) and the Electric Power Research Institute (EPRI) have developed a collaboration to conduct coordinated research to identify and address the technological challenges and opportunities that likely would affect the safe and economic operation of the existing NPP fleet over the postulated long-term time horizons. In this paper we describe a framework for developing and implementing a Risk-Informed Safety Margin Characterization (RISMC) approach to evaluate and manage changes in plant safety margins over long time horizons.
Date: July 1, 2009
Creator: Dinh, Nam & Szilard, Ronaldo
Partner: UNT Libraries Government Documents Department

Effect of in-pile degradation of the meat thermal conductivity on the maximum temperature of the plate-type U-Mo dispersion fuels

Description: Effect of in-pile degradation of thermal conductivity on the maximum temperature of the plate-type research reactor fuels has been assessed using the steady-state heat conduction equation and assuming convection cooling. It was found that due to very low meat thickness, characteristic for this type of fuel, the effect of thermal conductivity degradation on the maximum fuel temperature is minor. For example, the fuel plate featuring 0.635 mm thick meat operating at heat flux of 600 W/cm2 would experience only a 20oC temperature rise if the meat thermal conductivity degrades from 0.8 W/cm-s to 0.3 W/cm-s. While degradation of meat thermal conductivity in dispersion-type U-Mo fuel can be very substantial due to formation of interaction layer between the particles and the matrix, and development of fission gas filled porosity, this simple analysis demonstrates that this phenomenon is unlikely to significantly affect the temperature-based safety margin of the fuel during normal operation.
Date: November 1, 2009
Creator: Medvedev, Pavel G.
Partner: UNT Libraries Government Documents Department

Instrumentation, Control, and Intelligent Systems

Description: Abundant and affordable energy is required for U.S. economic stability and national security. Advanced nuclear power plants offer the best near-term potential to generate abundant, affordable, and sustainable electricity and hydrogen without appreciable generation of greenhouse gases. To that end, Idaho National Laboratory (INL) has been charged with leading the revitalization of nuclear power in the U.S. The INL vision is to become the preeminent nuclear energy laboratory with synergistic, world-class, multi-program capabilities and partnerships by 2015. The vision focuses on four essential destinations: (1) Be the preeminent internationally-recognized nuclear energy research, development, and demonstration laboratory; (2) Be a major center for national security technology development and demonstration; (3) Be a multi-program national laboratory with world-class capabilities; (4) Foster academic, industry, government, and international collaborations to produce the needed investment, programs, and expertise. Crucial to that effort is the inclusion of research in advanced instrumentation, control, and intelligent systems (ICIS) for use in current and advanced power and energy security systems to enable increased performance, reliability, security, and safety. For nuclear energy plants, ICIS will extend the lifetime of power plant systems, increase performance and power output, and ensure reliable operation within the system's safety margin; for national security applications, ICIS will enable increased protection of our nation's critical infrastructure. In general, ICIS will cost-effectively increase performance for all energy security systems.
Date: September 1, 2005
Partner: UNT Libraries Government Documents Department

Methods for quantifying uncertainty in fast reactor analyses.

Description: Liquid-metal-cooled fast reactors in the form of sodium-cooled fast reactors have been successfully built and tested in the U.S. and throughout the world. However, no fast reactor has operated in the U.S. for nearly fourteen years. More importantly, the U.S. has not constructed a fast reactor in nearly 30 years. In addition to reestablishing the necessary industrial infrastructure, the development, testing, and licensing of a new, advanced fast reactor concept will likely require a significant base technology program that will rely more heavily on modeling and simulation than has been done in the past. The ability to quantify uncertainty in modeling and simulations will be an important part of any experimental program and can provide added confidence that established design limits and safety margins are appropriate. In addition, there is an increasing demand from the nuclear industry for best-estimate analysis methods to provide confidence bounds along with their results. The ability to quantify uncertainty will be an important component of modeling that is used to support design, testing, and experimental programs. Three avenues of UQ investigation are proposed. Two relatively new approaches are described which can be directly coupled to simulation codes currently being developed under the Advanced Simulation and Modeling program within the Reactor Campaign. A third approach, based on robust Monte Carlo methods, can be used in conjunction with existing reactor analysis codes as a means of verification and validation of the more detailed approaches.
Date: April 7, 2008
Creator: Fanning, T. H. & Fischer, P. F.
Partner: UNT Libraries Government Documents Department

SHIPMENT OF NON-TRADITIONAL CONTENTS IN THE 9977 TYPE B PACKAGE

Description: The 9977 is a certified Type B Packaging authorized to ship uranium and plutonium in metal and oxide forms. These materials are typically confined within metallic containers designed for ease of handling and to prevent the spread of contamination. The Pacific Northwest National Laboratory (PNNL) uses Pu and U sources for the training of domestic and international customs agents in the identification and detection of radioactive materials (RAM). These materials are packed in polycarbonate containers which permit the trainees to view the RAM. The safety basis was made to authorize the use of these unusual containers. The inclusion of the PNNL Training Source Contents into the 9977 Packaging imposed unique conditions previously unanalyzed. The use of polycarbonate as a content container material, while different from any configuration previously considered, does not raise any safety issues with the package which continues to operate with a large safety margin for temperatures, pressures, containment, dose rates, and subcriticality.
Date: June 6, 2011
Creator: Abramczyk, G.; Loftin, B.; Bellamy, S. & Nathan, S.
Partner: UNT Libraries Government Documents Department

Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation.

Description: An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.
Date: March 1, 2010
Creator: Tentner, A. M.; Parma, E.; Wei, T.; Wigeland, R.; Division, Nuclear Engineering; SNL et al.
Partner: UNT Libraries Government Documents Department

Risk-Informed Safety Margin Characterization (RISMC): Integrated Treatment of Aleatory and Epistemic Uncertainty in Safety Analysis

Description: The concept of “margin” has a long history in nuclear licensing and in the codification of good engineering practices. However, some traditional applications of “margin” have been carried out for surrogate scenarios (such as design basis scenarios), without regard to the actual frequencies of those scenarios, and have been carried out with in a systematically conservative fashion. This means that the effectiveness of the application of the margin concept is determined in part by the original choice of surrogates, and is limited in any case by the degree of conservatism imposed on the evaluation. In the RISMC project, which is part of the Department of Energy’s “Light Water Reactor Sustainability Program” (LWRSP), we are developing a risk-informed characterization of safety margin. Beginning with the traditional discussion of “margin” in terms of a “load” (a physical challenge to system or component function) and a “capacity” (the capability of that system or component to accommodate the challenge), we are developing the capability to characterize probabilistic load and capacity spectra, reflecting both aleatory and epistemic uncertainty in system response. For example, the probabilistic load spectrum will reflect the frequency of challenges of a particular severity. Such a characterization is required if decision-making is to be informed optimally. However, in order to enable the quantification of probabilistic load spectra, existing analysis capability needs to be extended. Accordingly, the INL is working on a next-generation safety analysis capability whose design will allow for much more efficient parameter uncertainty analysis, and will enable a much better integration of reliability-related and phenomenology-related aspects of margin.
Date: October 1, 2010
Creator: Youngblood, R. W.
Partner: UNT Libraries Government Documents Department